IR 05000382/2016008

Nuclear Regulatory Commission Problem Identification and Resolution Inspection Report 05000382/2016008

ML17026A338
Person / Time
Site:	Waterford
Issue date:	01/26/2017
From:	Thomas Hipschman Division of Reactor Safety IV
To:	Chisum M Entergy Operations
References
IR 2016008
Download: ML17026A338 (37)

Text

January 26, 2017

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NUCLEAR REGULATORY COMMISSION PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000382/2016008

Dear Mr. Chisum:

On December 15, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a problem identification and resolution inspection at your Waterford Steam Electric Station, Unit 3, and discussed the results of the inspection with you and members of your staff. The results of this inspection are documented in the enclosed report.

The NRC inspection team reviewed the stations corrective action program and the stations implementation of the program to evaluate its effectiveness in identifying, prioritizing, evaluating, and correcting problems, and to confirm that the station was complying with NRC regulations and licensee standards for corrective action programs. Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety. However, the team noted that station personnel were challenged in several instances during the inspection period with effectively and timely completing actions to correct some conditions adverse to quality. These challenges are further discussed in the attached report.

The team also evaluated the stations processes for use of industry and NRC operating experience information and the effectiveness of the stations audits and self-assessments.

Based on the samples reviewed, the team determined that your staffs performance in each of these areas adequately supported nuclear safety.

Finally, the team reviewed the stations programs to establish and maintain a safety-conscious work environment, and interviewed station personnel to evaluate the effectiveness of these programs. Based on the teams observations and the results of these interviews, the team found no evidence of challenges to your organizations safety-conscious work environment.

Your employees appeared willing to raise nuclear safety concerns through at least one of the several means available.

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD ARLINGTON, TX 76011-4511 NRC inspectors documented three findings of very low safety significance (Green) in this report, each of which involved a violation of NRC requirements. Additionally, NRC inspectors documented one Severity Level IV violation with no associated finding. The NRC is treating all of these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at the Waterford Steam Electric Station.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at the Waterford Steam Electric Station.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Thomas R. Hipschman, Team Leader Inspection Programs and Assessment Team Division of Reactor Safety

Docket No. 50-382 License No. NPF-38

Enclosure:

Inspection Report 05000382/2016008 w/Attachments:

1. Supplemental Information 2. Information Request 3. Supplemental Information Request

ML17026A338 SUNSI Review By: ERuesch ADAMS Yes No Publicly Available Non-Publicly Available Non-Sensitive Sensitive Keyword:

NRC-002 OFFICE DRS/OB NRR/DIRS/PAB DRP/PBD DRS/EB1 DRP/PBD DRS/IPAT DRS/IPAT NAME MHayes DMerzke FRamirez GGeorge GMiller ERuesch THipschman SIGNATURE

/RA/

E-mail E-mail E-mail

/RA/

E-mail E-mail /RA/

DATE 1/18/17 1/19/17 1/19/17 1/23/17 1/25/17 1/26/17 1/26/17

Enclosure U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket(s):

05000382 License:

NPF-38 Report:

05000382/2016008 Licensee:

Entergy Operations, Inc.

Facility:

Waterford Steam Electric Station, Unit 3 Location:

17265 River Road Killona, LA 70057 Dates:

November 28 through December 15, 2016 Team Lead:

E. Ruesch, J.D., Senior Reactor Inspector Inspectors:

F. Ramirez, Senior Resident Inspector M. Hayes, Operations Engineer D. Merzke, Senior Reactor Operations Engineer G. George, Senior Reactor Inspector Approved By:

T. Hipschman, Team Leader Inspection Programs and Assessment Team Division of Reactor Safety

SUMMARY

IR 05000382/2016008; 06/06/2014 - 12/15/2016; WATERFORD 3; Problem Identification and

Resolution (Biennial)

The inspection activities described in this report were performed between November 28 and December 15, 2016, by three inspectors from the NRCs Region IV office, one inspector from the Office of Nuclear Reactor Regulation, and the resident inspector at the Waterford Steam Electric Station, Unit 3. The report documents three findings of very low safety significance (Green), each of which involved a violation of NRC requirements. Additionally, NRC inspectors documented in this report one Severity Level IV violation with no associated finding and one licensee-identified violation of very low safety significance. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.

Assessment of Problem Identification and Resolution

Based on its inspection sample, the team concluded that the licensee maintained a corrective action program in which individuals generally identified issues at an appropriately low threshold.

Once entered into the corrective action program, the licensee generally evaluated and addressed these issues appropriately and timely, commensurate with their safety significance.

The licensees corrective actions were generally effective, addressing the causes and extents of condition of problems. However, the team noted that station personnel were challenged in several instances during the inspection period with effectively and timely completing actions to correct some conditions adverse to quality.

The licensee appropriately evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The licensee incorporated industry and internal operating experience in its root cause and apparent cause evaluations.

The licensee performed effective and self-critical nuclear oversight audits and self-assessments.

The licensee maintained an effective process to ensure significant findings from these audits and self-assessments were addressed.

The licensee maintained a safety-conscious work environment in which personnel were willing to raise nuclear safety concerns without fear of retaliation.

Cornerstone: Mitigating Systems

• Green/SL-IV. The team identified a Green non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, which occurred when the licensee failed to dedicate commercial-grade relays for use in safety-related applications. After receiving information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality assurance standards, the licensee failed to take appropriate steps to accept these commercial-grade relays as basic components. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710 and initiated actions to ensure compliance with quality assurance requirements.

The failure to dedicate commercial-grade relays used asor intended for use asbasic components (in safety-related applications) as required by plant procedures and by 10 CFR Part 21 was a performance deficiency. This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system, or component, and operability was maintained.

The finding has a conservative bias cross-cutting aspect in the human performance cross-cutting area because licensee personnel improperly rationalized the adequacy of the nonconforming components to perform their safety-related functions (H.14).

Because this performance deficiency was also a violation that impacted the regulatory process, in that the licensee accepted a change to plant design without appropriate evaluation and notification, it was also evaluated for traditional enforcement. The team determined that the violation was Severity Level IV because it was similar to several examples in Section 6.5.d of the NRC Enforcement Policy. (Section 4OA2.5.a)

Green.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, that occurred when the licensee failed on two occasions to perform an operability determination for a nonconforming condition affecting numerous safety-related components. Following receipt of information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality requirements, licensee personnel failed to perform an operability evaluation. Later, during a Part 21 evaluation for the potential defect, the evaluator noted that an operability determination was needed, but failed to initiate the appropriate processes. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710, declared the affected components operable, but degraded, and initiated actions to restore full qualification.

Failures to perform an operability determination following identification of a nonconforming condition as required by station procedures were two examples of a performance deficiency.

This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it did not represent the actual loss of function of any system or train. The finding has an identification cross-cutting aspect in the problem identification and resolution cross-cutting area because licensee personnel failed to recognize a nonconforming condition as a condition adverse to quality (P.1). (Section 4OA2.5.b)

Green.

The team identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all the components that were being monitored. As a result, sufficient controls were not in place to ensure that a corrective action to prevent recurrence could prevent future piping failures.

The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2016-07487. The licensee will restore compliance by addressing the discrepancies between the requirements of the reconstituted feedwater/emergency feedwater monitoring plan associated with steam generator replacement induced vibration and the vibration data routinely collected by plant personnel.

The team determined that the performance deficiency was more-than-minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not adequately monitor vibrations in six components of the feedwater and emergency feedwater systems such that vibration-induced piping degradation could be detected and the availability and reliability of these systems would be maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the team determined that the finding was of very low safety significance because the answer to all the screening questions was no. This finding has a resolution cross-cutting aspect in the area of problem identification and resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, corrective actions to prevent recurrence of an adverse condition were closed without the issue being fully resolved (P.3). (Section 4OA2.5.c)

Other Findings

and Violations

• SL-IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2),

Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a proposed change, test, or experiment that would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Specifically, the licensee departed from their approved CEFLASH-4A methodology to determine steam generator internal differential loads caused by a main steam line break to an unapproved TRANFLOW methodology. In response to this issue, the licensee entered the issue into the corrective action program as Condition Report CR-WF3-2016-07639 and initiated actions to prepare a new evaluation under current regulatory guidelines or to submit a license amendment request to the NRC.

The licensees failure to obtain a license amendment prior to implementing a change that resulted in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses, as required by 10 CFR 50.59(c)(2) was a violation. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function. Therefore, this violation was processed through traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more-than-minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more-than-minor example of a change in requirements in the NRC Enforcement Manual,

Appendix E, Minor Violations - Examples, dated September 9, 2013. In accordance with the NRC Enforcement Policy, the significance determination process was used to inform the significance of the failure to obtain a license amendment prior to implementing a proposed change. The departure from the original CEFLASH-4A method to the TRANFLOW method to determine differential loads on steam generator internal structures following a main steam line break event was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,

Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the issue would not result in the complete or partial loss of a support system that contributes to the likelihood of an initiating event, or result in the steam generators violating accident leakage performance criterion. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects. (Section 4OA5.b)

Licensee-Identified Violations

A violation of very low safety significance that was identified by the licensee has been reviewed by the team. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

4OA2 Problem Identification and Resolution

The team based the following conclusions on a sample of corrective action documents that were open during the assessment period, which ranged from June 6, 2014, to the end of the on-site portion of this inspection on December 15, 2016.

.1 Assessment of the Corrective Action Program Effectiveness

a. Inspection Scope

The team reviewed a sample of condition reports, including associated root cause analyses and apparent cause evaluations, from the approximately 17,000 that the licensee had initiated or closed between June 2014 and November 2016. The majority of these (more than 14,000) were lower-level condition reports that did not require formal evaluations, or that were closed to the licensees work management system. The inspection sample of approximately 158 condition reports focused on higher-significance condition reports for which the licensee evaluated and took actions to address the cause of the condition. In performing its review, the team evaluated whether the licensee had properly identified, characterized, and entered issues into the corrective action program, and whether the licensee had appropriately evaluated and resolved the issues in accordance with established programs, processes, and procedures. The team also reviewed these programs, processes, and procedures to determine if any issues existed that may impair their effectiveness.

The team reviewed a sample of performance metrics, system health reports, operability determinations, self-assessments, trending reports and metrics, and various other documents related to the licensees corrective action program. The team evaluated the licensees efforts in determining the scope of problems by reviewing selected logs, work orders, self-assessment results, audits, system health reports, action plans, and results from surveillance tests and preventive maintenance tasks. The team reviewed daily condition reports and attended the licensees screening and performance improvement review group meetings to assess the reporting threshold and prioritization efforts, and to observe the corrective action programs interfaces with the operability assessment and work control processes. The teams review included an evaluation of whether the licensee considered the full extent of cause and extent of condition for problems, as well as a review of how the licensee assessed generic implications and previous occurrences of issues. The team assessed the timeliness and effectiveness of corrective actions, completed or planned, and looked for additional examples of problems similar to those the licensee had previously addressed. The team conducted interviews with plant personnel to identify other processes that may exist where problems may be identified and addressed outside the corrective action program.

The team reviewed corrective action documents that addressed past NRC-identified violations to evaluate whether corrective actions addressed the issues described in the inspection reports. The team reviewed a sample of corrective actions closed to other corrective action documents to ensure that the ultimate corrective actions remained appropriate and timely. The team reviewed a sample of condition reports where the licensee had changed the significance level after initial classification to determine whether the level changes were in accordance with station procedure and that the conditions were appropriately addressed.

The team considered risk insights from both the NRCs and the Waterfords risk models to focus the sample selection and plant tours on risk-significant systems and components. The team focused a portion of its sample on the cable vault and switchgear ventilation (SVS) and emergency safeguard features actuation (ESF)systems, which the team selected for a five-year in-depth review. The team conducted walk-downs of these systems and other plant areas to assess whether licensee personnel identified problems at a low threshold and entered them into the corrective action program.

b. Assessments

1. Effectiveness of Problem Identification

During the 30-month inspection period, licensee staff generated approximately 17,000 condition reports. The team determined that most conditions that required generation of a condition report by EN-LI-102, Corrective Action Program, and its progeny procedures had been appropriately entered into the corrective action program.

The team further noted that during 2016, following the implementation of a significant revision to the corrective action program, the number of condition reports closed to the work management system had steadily increased from approximately 300 to 500 per month, while the number of condition reports being assigned low-level corrective actions had decreased slightly. Based on discussions with licensee personnel at several levels of the organization, the team determined that these trends were due to a lower threshold for initiating condition reports, an improved understanding of the corrective action process by licensee personnel, and clearer guidance on what constitutes an adverse condition.

Overall, the team concluded that the licensee generally maintained a low threshold for the formal identification of problems and entry into the corrective action program for evaluation. Licensee personnel initiated an average of over 500 condition reports per month during the inspection period. All of the personnel interviewed by the team understood the requirements for condition report initiation; most expressed a willingness to enter newly identified issues into the corrective action program at a very low threshold.

2. Effectiveness of Prioritization and Evaluation of Issues

The sample of condition reports reviewed by the team focused primarily on issues screened by the licensee as having higher-level significance, including those that received cause evaluations, those classified as significant conditions adverse to quality, and those that required engineering evaluations. The team also reviewed a number of condition reports that included or should have included immediate operability determinations to assess the quality, timeliness, and prioritization of these determinations.

The team identified two examples where the licensee did not thoroughly evaluate adverse conditions, which in one case led to multiple failures to take required actions:

• The team reviewed the root cause evaluation for Condition Report CR-WF3-2015-4094, Un-isolable Feedwater Drain Line Leak, and noted that the scope of the extent of condition was narrow. For this event, the licensee found an un-isolable, through-wall fatigue crack in the steam generator number 1 main feedwater regulating valve downstream drain isolation, which required a forced shutdown. The licensees extent of condition review focused on components in the feedwater system that had a similar configuration as the one that failed (close proximity welds, cantilevered design, and no supports). The team noted that the licensee could have extended the review to safety-related systems where drain valves or other components had the same configuration that caused the un-isolable leak to determine if the condition exists. The licensee documented this issue in Condition Report CR-WF3-2016-07463.

• In May 2016, the licensee initiated Condition Report CR-WF3-2016-03525 to evaluate a Part 21 report from Electroswitch, which documented a nonconforming condition potentially affecting more than 124 relays installed in safety-related applications. The licensee assigned an operability code of ADMIN NA for the nonconforming condition, bypassing the operability determination process. In June 2016, an engineer performed a Part 21 screening of the condition using EN-LI-108-01, Attachment 9.1. The engineer noted that the condition should be rescreened for operability and reportability, but failed to initiate a new condition report to perform these evaluations.

These two examples of failures to perform required operability determinations are documented as a finding in Section 4OA2.5.b below. Further, these evaluation failures contributed to the licensees failure to control nonconforming quality parts as required by regulations, which is described as a finding in Section 4OA2.5.a below.

Overall, the team determined that the licensees process for screening and prioritizing issues that had been entered into the corrective action program supported nuclear safety. The licensees operability determinations were for the most part consistent, accurately documented, and completed in accordance with procedures.

3. Effectiveness of Corrective Actions

In general, the corrective actions identified by the licensee to address adverse conditions were effective. However, the team noted a number of instances in which corrective actions had been untimely or incompletely accomplished:

• In April 2016, the NRC identified that the licensee was not monitoring the vibrations for the cable vault and switchgear ventilation system safety-related fans at the frequency required by Waterfords preventive maintenance template. At the time, it was identified that even though the system components were classified appropriately, the preventive maintenance tasks associated with the fans had not been identified as such. As a result, the licensee was taking fan vibration data every 18 months instead of every three months. When reviewing the corrective actions associated with this issue, which were documented in Condition Report CR-WF3-2016-04544, the team noted that Corrective Action 13 was created to verify that the preventive maintenance tasks associated with all the components in the cable vault and switchgear ventilation system were being performed in accordance with the preventive maintenance template. The team noted that the corrective action was initially due in November 24, 2016, but that it was extended to January 27, 2017. The team also noted that several of the preventive maintenance tasks that would be required per the sites preventive maintenance template, would have been changed to a more frequent basis, and would have been completed within the window of the due date extension period. As a result of the teams questions, the licensee reviewed the condition. The licensee determined that the preventive maintenance tasks in question had an adequate frequency and only required a documented basis for the deviation from the preventive maintenance template. The failure to pre-plan and perform preventative maintenance on safety-related components in accordance with EN-DC-335, Preventative Maintenance Basis Template, was a minor performance deficiency because the team determined that the documentation issue would not have had an impact on equipment performance. The licensee documented this issue in Condition Report CR-WF3-2016-07790.

• In 2014, the licensee initiated an apparent cause evaluation under Condition Report CR-WF3-2014-04930 to determine the cause of corrosion of heat exchanger tubes in the dry cooling tower, the stations ultimate heat sink.

This apparent cause evaluation was revised several times, resulting in numerous cross-references and a corrective action plan containing over 50 tracked actions. Two of these corrective actions (14 and 50) were to submit an action request to develop preventive maintenance (PM) tasks to clean the coils and for the system engineer to inspect their condition following cleaning. After development of these tasks, a system engineer inadvertently changed the PM frequency to 18 years instead of the intended 24 months.

When this was identified following the teams questions, the PM task was past its 24-month due date, but was still within the allowed grace period.

Cleaning had been accomplished under a different PM task, which did not include the additional inspection tasks added to correct the problems that were the original subject of the CR 14-4930 ACEno engineering inspection was documented. This failure to develop and maintain appropriate preventive maintenance schedules constitutes a minor violation of Technical Specification 6.8.1 that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The licensee documented this performance deficiency in Condition Report CR-WF3-2016-07454.

• Condition Report CR-WF3-2015-07994 documented 23 condition reports over approximately 14 years identifying ongoing degradation (corrosion) of safety-related motor control centers in the dry cooling tower areas.

Additionally, two roll-up condition reports identified this condition, and licensee management designated it as a station top ten reliability issue.

As of November 2015, the condition adverse to quality had still not been corrected, as documented in Condition Report CR-WF3-2015-07977.

The team noted that the licensee self-identified this performance deficiency and has developed a corrective action plan that appears adequate to correct the condition; the licensee currently plans to perform corrective maintenance during the upcoming refueling outage. This performance deficiency is a licensee-identified violation of very low safety significance, further described in Section 4OA7 below.

• In 2015, following the identification of multiple pipe support failures resulting from excessive feed water piping vibrations, the licensee initiated a root cause evaluation under Condition Report CR-WF3-2015-04094. In this root cause evaluation, the licensee determined that the required corrective action to preclude repetition (CAPR) was to develop a plan to monitor the feed water piping vibrations. The licensee identified monitoring points and then closed the CAPR to this plan. The team reviewed the licensees monitoring plan and noted that it provided no acceptance criteria for six of the monitoring points, two of which were on safety-related emergency feedwater piping.

This failure to provide acceptance criteria, contrary to quality assurance program requirement, is documented as a finding in Section 4OA2.5.c below.

Overall, the team concluded that despite these examples, the licensee generally identified effective corrective actions for the problems evaluated in the corrective action program. The licensee generally implemented these corrective actions in a timely manner, commensurate with their safety significance, and reviewed the effectiveness of the corrective actions appropriately.

.2 Assessment of the Use of Operating Experience

a. Inspection Scope

The team examined the licensees program for reviewing industry operating experience, including reviewing the governing procedures. The team reviewed a sample of industry operating experience communications and the associated site evaluations to assess whether the licensee had appropriately assessed the communications for relevance to the facility. The team also reviewed assigned actions to determine whether they were appropriate. Attachment 1 includes a list of documents the team reviewed in performing its assessment.

b. Assessment

Overall, the team determined that the licensee appropriately evaluated industry operating experience for its relevance to the facility. The team noted that the licensee appropriately evaluated industry operating experience when performing root cause analysis and apparent cause evaluations. The licensee appropriately incorporated both internal and external operating experience into lessons learned for training and pre-job briefs.

.3 Assessment of Self-Assessments and Audits

a. Inspection Scope

The team reviewed a sample of licensee self-assessments and audits to assess whether the licensee was regularly identifying performance trends and effectively addressing them. The team also reviewed audit reports to assess the effectiveness of assessments in specific areas. The specific self-assessment documents and audits reviewed are listed in Attachment 1.

b. Assessment

Overall, the team concluded that the licensee had an effective self-assessment and audit process. The team determined that self-assessments were self-critical and thorough enough to identify deficiencies.

1. Inspection Scope

The team interviewed 30 individuals in five focus groups. The purpose of these interviews was

(1) to evaluate the willingness of licensee staff to raise nuclear safety issues, either by initiating a condition report or by another method,

(2) to evaluate the perceived effectiveness of the corrective action program at resolving identified problems, and

(3) to evaluate the licensees safety-conscious work environment. The focus group participants included personnel from security, radiation protection, chemistry, engineering, operations, production, maintenance, and programs. At the teams request, the licensees regulatory affairs staff selected the participants blindly from these work groups, based partially on availability. To supplement these focus group discussions, the team interviewed the employee concerns program manager to assess her perception of the site employees willingness to raise nuclear safety concerns. The team reviewed the employee concerns program case log and select case files. The team also reviewed the minutes from the licensees most recent safety culture monitoring panel meetings.

1. Willingness to Raise Nuclear Safety Issues

All individuals interviewed indicated that they would raise nuclear safety concerns.

All felt that their management was receptive to nuclear safety concerns and was willing to address them promptly. All of the interviewees further stated that if they were not satisfied with the response from their immediate supervisor, they had the ability to escalate the concern to a higher organizational level. Most expressed positive experiences after raising issues to their supervisors. All expressed positive experiences documenting most issues in condition reports.

2. Employee Concerns Program

All interviewees were aware of the employee concerns program. Most explained that they had heard about the program through various means, such as posters, training, presentations, and discussion by supervisors or management at meetings. Most interviewees stated that they would use the employee concerns program if they felt it was necessary. All but one expressed confidence that their confidentiality would be maintained if they brought issues to Employee Concerns.

3. Preventing or Mitigating Perceptions of Retaliation

When asked if there have been any instances where individuals experienced retaliation or other negative reaction for raising issues, all individuals interviewed stated that they had neither experienced nor heard of an instance of retaliation, harassment, intimidation, or discrimination at the site. The team determined that processes in place to mitigate these issues were being successfully implemented.

.5 Findings

a. Failure to Control Nonconforming Parts

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XV, which occurred when the licensee failed to dedicate commercial-grade relays for use in safety-related applications. After receiving information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality assurance standards, the licensee failed to take appropriate steps to accept these commercial-grade relays as basic components.

Description.

On May 10, 2016, Electroswitch submitted a Part 21 report to the NRC (ML16139A834) documenting that a large population of rotary switches and relays it had manufactured since 1984 had not conformed to quality assurance program requirements. (Electroswitch has discontinued its 10 CFR Part 50, Appendix B, quality assurance program effective March 24, 2016.) The Part 21 notification noted that Electroswitch did not procure materials, parts, equipment and/or services from an Appendix B supplier nor were applicable Commercial Grade Surveys, Source Inspections and Material Analyses performed for a number of materials used in the manufacture of these components, which it sold as safety-related basic components Waterford 3 was listed as an affected facility in a May 11, 2016, update to Electroswitchs report.

On May 25, 2016, the licensee initiated Condition Report CR-WF3-2016-03525 to evaluate Electroswitchs Part 21 report. On June 13, 2016, the licensee performed a Part 21 screening using EN-LI-108-01 Attachment 9.1, which required that for safety-related parts installed in the plant, personnel performing the screen assure that the issue has been or is being evaluated under 10 CFR 50.72 and/or 50.73, and includes the requirement of 10 CFR Part 21. Issue a new CR if needed. The evaluator noted that, Electroswitch has determined it does not have the capability to perform the evaluation to determine if a defect, which could create a substantial safety hazard exists. On June 21, a Waterford Steam Electric Station licensing engineer documented that no Part 21 notification was required because a Part 21 notification had been issued by the vendor. However, the engineer failed to account for Electroswitchs statement that it lacked information to determine whether the identified deviation could create a substantial safety hazard, and was therefore a defect, at any individual station.

On July 19, 2016, the licensee completed an engineering evaluation, inappropriately concluding the following:

Electroswitch relays and rotary switches have been designed by Electroswitch for decades and have been proven through testing and performance (i.e., OEs) as being reliable. Based on corrective actions taken by Electroswitch and the applications and reliability of these relays and switches, it is concluded that the non-conformance issues on Electroswitch products described in this 10 CFR Part 21 report does not create a substantial safety hazard such that the loss of a safety-related function is initiated. Therefore, the function of a Technical Specification system, structure, or component is not adversely affected as a result of this 10 CFR Part 21 report.

The licensee evaluator further concluded that, No additional corrective actions are required and Electroswitch materials in the warehouse may be issued as required.

These conclusions accepted nonconforming components as fully qualified and permitted further installation of commercial grade components in safety-related applications with no acceptance testing or other dedication, contrary to the requirements of 10 CFR Part 21.1

Following identification of this condition by the team on November 30, 2016, the licensee initiated Engineering Change EC-68416, which repeated the conclusions of the July 19 evaluation verbatim. The evaluator again concluded that the installation of nonconforming or commercial-grade components in safety-related applications as fully qualified basic components was acceptable, with no dedication or further evaluation required. On December 13, 2016, after discussion with the team, licensee personnel initiated Condition Report CR-WF3-2016-07710. Under this condition report, the licensee determined the affected components to be operable, but degraded, and initiated further actions to perform appropriate acceptance testing or dedication to restore full qualification.

Analysis.

The failure to dedicate commercial-grade relays used asor intended for use asbasic components (in safety-related applications) as required by plant procedures and by 10 CFR Part 21 was a performance deficiency. This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system, or component, and operability was maintained. The finding has a conservative bias cross-cutting aspect in the human performance cross-cutting area because licensee personnel improperly

1 On June 24, 2016, Electroswitch completed additional testing of the original test specimens from 1984-85, using the acceptance criteria of the appropriate quality standards. The specimen components passed these tests. However, because Electroswitch had discontinued its Appendix B quality assurance program effective March 24, 2016, these tests were not performed under an approved quality assurance program. Thus, they could not on their own demonstrate qualification of the nonconforming components.

rationalized the adequacy of the nonconforming components to perform their safety-related functions (H.14).

Because this performance deficiency was also a violation that impacted the regulatory process, in that the licensee accepted a change to plant design without appropriate evaluation and notification, it was also evaluated for traditional enforcement. The team determined that the violation was Severity Level IV because it was similar to several examples in Section 6.5.d of the NRC Enforcement Policy.

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion XV, requires that nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures. Contrary to this requirement, in May 2016, the licensee failed to review and accept, reject, repair or rework nonconforming items in accordance with documented procedures. Specifically, after receipt of information from a vendor that a number of relays did not conform to quality requirements, licensee engineers cited performance history to justify leaving the nonconforming relays in safety-related service, and failed to dedicate these commercial-grade parts as described in 10 CFR Part 21. Because this violation is of very low safety significance and Severity Level IV, and was entered into the licensees corrective action program (CR-WF3-2016-07710), it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Manual: NCV 05000382/2016008-01, Failure to Control Nonconforming Parts.

b. Failure to Perform Operability Determinations for Nonconforming Conditions

Introduction.

The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, that occurred when the licensee failed on two occasions to perform an operability determination for a nonconforming condition affecting numerous safety-related components. Following receipt of information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality requirements, licensee personnel failed to perform an operability evaluation.

Later, during a Part 21 evaluation for the potential defect, the evaluator noted that an operability determination was needed, but failed to initiate the appropriate processes.

Description.

In May 2016, the licensee initiated Condition Report CR-WF3-2016-03525 to evaluate a Part 21 report from Electroswitch, which documented a nonconforming condition potentially affecting more than 124 relays installed in safety-related applications. The licensee assigned an operability code of ADMIN NA for the nonconforming condition, bypassing the operability determination process. Per Step 5.3 of EN-OP-104, Revision 10, which was in effect at the time, an operability evaluation was required for this nonconforming condition.

In July 2016, an engineer performed a Part 21 screening of the condition using EN-LI-108-01, Attachment 9.1. The engineer noted that the condition should be rescreened for operability and reportability. Initiation of a new condition report was required by procedure. Further, in the initial condition report, operators noted that after initial Part 21 and operating experience evaluations were performed, If either determines that a degraded or nonconforming condition exists, then an additional Condition Report will be generated to evaluate the specific OPERABILITY of any potentially affected components. However, no new condition report was initiated and no operability determination was performed.

Analysis.

Failures to perform an operability determination following identification of a nonconforming condition as required by station procedures were two examples of a performance deficiency. This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was did not represent the actual loss of function of any system or train. The finding has an identification cross-cutting aspect in the problem identification and resolution cross-cutting area because licensee personnel failed to recognize a nonconforming condition as a condition adverse to quality (P.1).

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, Nonconforming Materials, Parts, or Components, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this requirement, in May and June 2016, the licensee failed to accomplish activities affecting quality in accordance with prescribed instructions and procedures. Specifically, licensee personnel failed to perform operability evaluations on nonconforming components as required by Procedure EN-OP-104, Operability Determination Process. Because this violation is of very low safety significance and Severity Level IV, and was entered into the licensees corrective action program (CR-WF3-2016-07710), it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Manual:

NCV 05000382/2016008-02, Failure to Perform Operability Determinations for Nonconforming Conditions.

c. Failure to Include Appropriate Quantitative Acceptance Criteria for the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement-Induced Vibration

Introduction.

The team identified a finding of very low safety significance and associated Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all the components that were being monitored.

Description.

On June 22, 2015, the licensee shutdown the plant to repair a steam leak that was discovered near the steam generator number 1 main feedwater regulation valve downstream drain isolation (FW-174A). The licensee determined, through their cause evaluation process, that the root cause of the steam leak near FW-174A was that insufficient rigor in the Feedwater Piping Vibration Monitoring Plan failed to recognize a latent vulnerability in the post steam generator replacement environment. This Feedwater Piping Vibration Monitoring Plan was created and instituted following a January 21, 2013, automatic reactor trip that occurred following the startup from Refueling Outage 18, when both steam generators were replaced. At the time, the licensee determined that the January 2013 automatic reactor trip occurred due to feedwater system vibrations that were induced by the steam generator replacement.

Following the June 22, 2015, plant shutdown, the licensee concluded that the monitoring plan for feedwater vibration was not sufficiently rigorous in its development and implementation, lacked ownership, and did not account for critical attributes such as insulation or interference dampening when decisions were made to remove points from the plan. To address the deficiencies with the initial monitoring plan, the licensee created a corrective action to prevent recurrence that would implement a reconstituted vibration monitoring plan. The Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration would expand the scope of the monitored points in the feedwater and emergency feedwater systems, would account for multiple factors such as the configuration of the monitored points, would establish more stringent criteria, and would implement increased frequency on trending requirements.

The team reviewed the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration, dated December 17, 2015, and noted that 11 of the 28 monitoring points did not include a range for acceptable vibration levels. Five of the monitoring points were for feedwater system components located inside containment. The licensee inspected these components during Refueling Outage 20 and concluded based on satisfactory results, that continuous monitoring inside containment was not necessary. However, six monitoring points that included components in the safety-related emergency feedwater system did not have acceptance criteria. In addition, as a result of the teams questions, the licensee identified that the predictive maintenance data routinely collected by plant personnel to monitor feedwater and emergency feedwater system did not match the reconstituted monitoring plan. Therefore, theteam concluded that the licensee closed the corrective action to prevent recurrence from the June 22, 2015, feedwater system steam leak without the issue being fully resolved.

Analysis.

The team determined that the licensees failure to develop acceptance criteria for all the components in the reconstituted feedwater/emergency feedwater monitoring plan associated with steam generator replacement-induced vibration was a performance deficiency. The team concluded that the performance deficiency was more-than-minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not adequately monitor vibrations for six components in the feedwater and emergency feedwater systems such that vibration-induced piping degradation could be detected and the availability and reliability of these systems would be maintained. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was did not represent the actual loss of function of any system or train. This finding has a resolution cross-cutting aspect in the area of problem identification and resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, corrective actions to prevent recurrence of an adverse condition were closed without the issue being fully resolved (P.3).

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to the above, since December 17, 2015, the licensee did not include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration, which was created as part of a corrective action to prevent recurrence of vibration-induced piping failures, did not include an acceptable range for vibration levels for all the components that were being monitored.

As a result, sufficient controls were not in place to ensure that a corrective action to prevent recurrence could prevent future piping failures. The licensee will restore compliance by addressing the discrepancies between the requirements of the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration and the vibration data routinely collected by plant personnel. Because this violation was of very low safety significance and the licensee entered the issue into their corrective action program as Condition Report CR-WF3-2016-07487, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000382/2016008-03, Failure to Include Appropriate Quantitative Acceptance Criteria for the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration.

4OA5 Other Activities

.1 (Closed) Unresolved Item URI 05000382/2014008-06, Feedwater System

Vibrations

a. Scope

and Discussion

The team reviewed information associated with Unresolved Item 05000382/2014008-06, Feedwater System Vibrations, written to document an issue of concern that vibration-induced failures of main feedwater and emergency feedwater components could be a result inadequate design reviews. The team compared documented actions taken by the licensee after the onset of vibration to the requirements of the applicable design standards for the Quality Group B and C portions of the main feedwater and emergency feedwater components. The team reviewed pipe support vendor documentation to determine if pipe supports would remain qualified when subjected to increased vibration.

During plant startup on January 21, 2013, and a subsequent plant trip, the licensee failed to monitor pipe vibration of the main feedwater system in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and American National Standards Institute B31.1 design standards. In the subsequent plant startup, the licensee took appropriate corrective action to monitor vibration and take corrective actions. The NRC had previously issued a non-cited violation in NRC Inspection Report 0500382/2013003 for the licensees failure to monitor vibration during plant startup.

On May 22, 2013, the licensee documented Condition Report CR-WF3-2013-00445, Corrective Action 82, which documented the licensees feasibility study that reviewed main feedwater system mechanical snubbers that could be affected by increased vibration and the recommendation to provide additional supports to dampen the main feedwater system. The licensee recommended, at the time, that additional supports were not necessary. The licensee recommended that mechanical snubbers would be replaced, beginning in the next refueling outage, with hydraulic snubbers because the hydraulic snubbers have a longer service life in high vibration areas and could dampen some vibration. Engineering change packages were generated to replace hydraulic snubbers in the spring 2014 refueling outage. Additionally, a recommendation was made to test 100 percent of snubbers on the main feedwater system at every refueling outage as part of the normal snubber testing program to identify and manage snubber failures. At present date, all important-to-safety main feedwater mechanical snubbers in containment have been replaced with hydraulic snubbers.

Based on this review, the team determined that the licensees actions to date met the requirements for monitoring and correcting vibration within the applicable design standards. Additionally, the corrective actions for the previously identified non-cited violation were appropriate. Therefore, this unresolved item was closed.

b. Findings

Departure from Approved Method to Determine Steam Generator Internal Loads During Main Steam Line Break

Introduction.

The team identified a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2), Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a proposed change, test, or experiment that would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Specifically, the licensee departed from their approved CEFLASH-4A methodology to determine steam generator internal differential loads caused by a main steam line break to an unapproved TRANFLOW methodology.

Description.

On December 4, 2012, the licensee approved 10 CFR 50.59, Evaluation 2012-03. The evaluated change was to incorporate five computer codes into the updated final safety analysis report that were used for evaluating the structural integrity of the Waterford 3 replacement steam generators. One of the codes used was the TRANFLOW computer code. The licensees vendor, Westinghouse, used this computer code to determine the pressure drop across the replacement steam generator secondary side internal components. As described in the Waterford 3 Updated Final Safety Analysis Report, Section 3.9.1.2.2.1.28, the original computer code used for determining pressure drops across the original steam generator secondary side components was CEFLASH-4A. The licensee determined that the change in computer codes from CEFLASH-4A to TRANFLOW did not result in departure from a method of evaluation described in the final safety analysis report used in the safety analysis. The basis for this conclusion was that the TRANFLOW code met the criteria of Section 4.3.8.1 of NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1.

Specifically, the licensee concluded that this computer code was not a departure from an approved method because this was a change to an element of the analysis methods that yielded conservative results, or results that are essentially the same.

The basis for this conclusion was Westinghouse LTR-NCE-05-145, TRANFLOW Computer Code Comparison to CEFLASH-4B Analysis of the Watts Bar Replacement Steam Generator during a Feedwater Line Break, dated October 14, 2005. This Westinghouse letter described the use of the TRANFLOW computer code. It discussed that TRANFLOW has not been approved by the NRC for determining loads on steam generator internals during a steam line break; however, it produces results comparable to NRC-approved codes that were used for that purpose.

Westinghouse LTR-NCE-05-145 included a comparison of TRANFLOW results to the approved CEFLASH-4B, which is an upgraded version of CEFLASH-4A. The licensee concluded that since the TRANFLOW results were similar to the CEFLASH-4B results, and CEFLASH-4B provides identical results to CEFLASH-4A, then TRANFLOW produces comparable results to the approved CEFLASH-4A.

Additional basis for the conclusion was Westinghouse LTR-NCE-04-28, Position Paper on the Use of the TRANFLO/TRANFLOW Computer Program in Steam Generator Design Analysis, dated May 13, 2004. This letter discussed that, although not explicitly approved by the NRC, this computer code had been used since the early 1970s for determining internal loads in steam generators. Additionally, the NRC requested Westinghouse rerun a Diablo Canyon steam generator analysis using an NRC-approved code that resulted in acceptable results comparable to TRANFLOW. The letter also discussed that since TRANFLOW was controlled by approved quality assurance program, the NRC implicitly approved the use of the code in safety analyses.

The licensee is permitted to make changes to the facility as described in the updated final safety analysis report without prior NRC approval, provided that these changes did not result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, states that the methods described in Nuclear Energy Institute NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1, are acceptable to the NRC staff for complying with the provisions of 10 CFR 50.59.

NEI 96-07, Section 4.3.8, provides the criteria for determining if an activity results in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses.

Section 4.3.8.1 provides the criteria for changing one or more elements of a method of evaluation. Section 4.3.8.2 provides the criteria for changing from one method of evaluation to another.

The licensee determined that the change to the TRANFLOW code was a change to one or more elements of the CEFLASH-4A code; therefore, the criteria of NEI 96-07, Section 4.3.8.1, applied to the change. Based on this criteria, the change would not be a departure from an approved method because the method yielded conservative results and were essentially the same. The team determined this conclusion and use of the criteria in Section 4.3.8.1 to be incorrect.

The team determined that the change to the TRANFLOW code from CEFLASH-4A code was a change from one method described in the final safety analysis report to another method; therefore, the criteria of NEI 96-07, Section 4.3.8.2, applied to the change.

Based on this criteria, the change to a new computer code would not be a departure from an approved method if the new method is considered to be approved by the NRC for the intended application. A new method is approved by the NRC for intended application if it is previously approved earlier by NRC through generic approval of analysis methodologies in a topical report or safety evaluation report to the methodology owner. Additionally, a new method is approved if the specific analysis was approved by the NRC through a specific plants licensing bases. The user of the new methodology would then use the method as approved by the NRC.

The team searched the licensees design bases, licensing bases, and the NRCs licensing database to determine if TRANFLOW had ever been approved by the NRC to determine the pressure drop across the replacement steam generator secondary side internal components. The search confirmed that the NRC had never approved the TRANFLOW code for this specific evaluation. This result was consistent with the information provided by Westinghouse letters previously discussed.

Analysis.

The licensees failure to obtain a license amendment prior to implementing a change that resulted in a departure from a method of evaluation described in the final safety analysis report(as updated) used in establishing the design bases or in the safety analyses, as required by 10 CFR 50.59(c)(2) was a violation. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function.

Therefore, this violation was processed through the traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more-than-minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more-than-minor example of a change in requirements in the NRC Enforcement Manual, Appendix E, Minor Violations

- Examples, dated September 9, 2013. In accordance with the NRC Enforcement Policy, the significance determination process was used to inform the significance of the failure to obtain a license amendment prior to implementing a proposed change. The departure from the original CEFLASH-4A method to the TRANFLOW method to determine differential loads on steam generator internal structures following a main steam line break event was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the issue would not result in the complete or partial loss of a support system that contributes to the likelihood of an initiating event, or result in the steam generators violating accident leakage performance criterion. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.

Traditional enforcement violations are not assessed for cross-cutting aspects.

Enforcement.

Title 10 CFR 50.59(c)(2), states, in part, a licensee shall obtain a license amendment prior to implementing a proposed change, test, or experiment if the change, test, or experiment wouldresult in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Contrary to this requirement, since December 12, 2012, the licensee failed to obtain a license amendment prior to implementing a change to the facility as described in the final safety analysis report that resulted in departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Specifically, the licensee failed to obtain a licensed amendment for a departure in the methodology to determine differential loads on steam generator internal structures following a main steam line break event, as stated in the Waterford 3 Updated Final Safety Analysis Report, Section 3.9.1.2.2.1.28. In response to this issue, the licensee entered the issue into the corrective action program as Condition Report CR-WF3-2016-07639 and initiated actions to prepare a new evaluation under current regulatory guidelines or to submit a license amendment request to the NRC. Because the associated finding was of very low safety significance and because the licensee entered it into its corrective action program, this violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000382/2016008-04, Departure from Approved Method to Determine Steam Generator Internal Loads During Main Steam Line Break.

.2 (Closed) Violation VIO 05000382/2013004-01, Failure to Make a Report Required

by 10 CFR 21.21

On November 20, 2013, the NRC issued a Notice of Violation to Entergy Operations, Inc., citing the licensees failure to complete an evaluation of a deviation in a basic component within 60 days of discovery at Waterford 3, as required by 10 CFR 21.21.

This Notice of Violation closed unresolved item URI 05000382/2009010-01, documented on November 5, 2009 (ML093100238). In its documentation of the URI, the NRC noted that the licensees procedures did not incorporate applicable NRC guidance on 10 CFR Part 21 reporting.

During the 2014 problem identification and resolution inspection, documented in NRC Inspection Report 2014008 (ML14206A858), the NRC reviewed the licensees corrective actions to restore compliance, including revisions to its Part 21 reporting procedure, EN-LI-108-01. The 2014 team determined that these procedure revisions included the addition of inaccurate interpretations of regulations and regulatory guidance that were likely to result in the licensees continuing failure to perform required evaluations or to make timely required reports. As a result, the licensees corrective actions were inadequate to ensure future compliance with 10 CFR Part 21 reporting requirements.

Entergy Operations, Inc., was therefore not in compliance with the requirement of 10 CFR 21.21(a) that it adopt appropriate procedures to... evaluate deviations and failures to comply... in order to identify a reportable defect or failures to comply that could create a substantial safety hazard. Based on this failure to restore compliance, the original violation remained open.

During this inspection, the team reviewed the most recent revision to EN-LI-108-01 (Revision 8), and determined that the licensee had deleted the inaccurate interpretations of regulations that had been included in earlier revisions. The licensee had thus restored compliance with § 21.21(a).

Violation VIO 05000382/2013004-01 and Enforcement Action EA-12-257 are closed.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On December 15, 2017, the team presented the inspection results to Mr. M. Chisum, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the team had been returned or destroyed.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.

• Title 10 CFR Part 50, Appendix B, Criterion XVI, requires that measures shall be established to ensure that conditions adverse to quality are promptly identified and corrected. Contrary to this requirement, prior to late 2015, the licensee failed to establish measures to ensure that a condition adverse to quality was promptly identified and corrected. Specifically, 23 condition reports over 14 years documented corrosion of safety-related motor control centers in the dry cooling tower areas. During that time, several work orders had been closed with no actions taken, and other actions had been unsuccessful in mitigating the corrosion. The licensee has since identified the condition and commenced actions to mitigate the corrosion, as described in Condition Reports CR-WF3-2015-7994, -7977, and -6072. This performance deficiency was of very low safety significance because it did not result in the loss of operability or functionality of any system or train.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

C. Becnel, Employee Concerns Coordinator

J. Crews, Assistant Maintenance Manager

D. Duplessis, EDG System Engineer

S. Fontenot, Manager Performance Improvement

R. Gendusa, Operating Experience Specialist

M. Groome, Senior Engineer, Safety Injection

J. Hoss, Senior Engineer in Design Engineering

D. Litolff, Senior Reactor Operator - Control Room Supervisor

T. Manziel, CAA Senior Specialist

S. Meiklejohn, Senior Licensing Specialist

M. Mills, Manager Nuclear Oversight, Quality Assurance

N. Petit, Engineer Supervisor in Design Engineering

G. Pickering, Engineer Supervisor in Design Engineering

J. Sanchez - Senior Maintenance Specialist - Mechanical Maintenance

G. Setoon - Coordinator Equipment Reliability

D. Veiner, Engineering Supervisor Design Engineering

NRC Personnel

C. Speer, Resident Inspector

D. Loveless, Senior Reactor Analyst

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000382/2016008-01 NCV Failure to Control Nonconforming Parts (Section 4OA2)

05000382/2016008-02 NCV Failure to Perform Operability Determinations for Nonconforming Conditions (Section 4OA2)

05000382/2016008-03 NCV Failure to Include Appropriate Quantitative Acceptance Criteria for the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration (Section 4OA2)

05000382/2016008-04 NCV Departure from Approved Method to Determine Steam Generator Internal Loads During Main Steam Line Break (Section 4OA5)

Closed

05000382/2014008-06 URI Feedwater System Vibrations (Section 4OA5)

05000382/2013004-01 VIO Failure to Make a Report Required by 10 CFR 21.21